The invention is directed to a molded body made of graphite and an inorganic binder for the safe, long time fixation of toxic and radioactive wastes and a process for its production.
Spent fuel elements from nuclear reactors after a certain intermediate storage time must be sent to a final disposal. For this purpose there are discussed two possibilities:
1. The reprocessing of the fuel element with return of the fuel material into the preparation of the fuel element as well as separation, conditioning and final storage of the fission products.
2. The direct storage of the spent fuel elements.
In both cases through a suitable embedding matrix which is especially corrosion and leach resistant or a corresponding container material there must be insured that the enclosed highly active waste remains in the storage place for a thousand years or longer and does not get into the biosphere.
The radioactive waste accumulating in the reprocessing of spent fuel elements must be brought into a form capable of final storage in order to be final stored. Hereby there is required for economical reason a high loading with waste. For this purpose there is needed already before the solidification step a great reduction in volume, for example through evaporation.
There are known several processes for the solidification of highly radioactive waste. For example there is first carried out a calcination of the waste in a fluidized bed between 350.degree. and 900.degree. C. Thereby there is obtained a mixture of oxides, which as powder or granulate is bound in a glass-like on ceramic matrix and therethrough is consolidated to a final storable product.
For the embedding of medium and low active wastes there are known processes according to which the waste material is heated, e.g. with bitumen, and is subjected to an extrusion process. Thereby the radioactive waste is embedded in the bitum matrix, filled hot into drums and final stored.
A further process is the fixation of radioactive waste in cement or concrete. Hereby the waste is processed customarily in the form of a salt concentrate or slurry, which is composed of about 70-80 weight % of liquid and 20-30 weight % of solid components. The slurry is mixed with cement and allowed to set. This process can be carried out directly in the final storage tanks.
Furthermore, there are known processes for the conditioning of radioactive wastes in which the waste is mixed into a resin which is preferably hardenable at room temperature, and then polymerized to a solid block.
These known processes possess a series of disadvantages, especially for the higher activity concentrations. Thus the nitrification of the wastes is carried out at high temperatures, usually above 1000.degree. C. At this temperature several salts are already volatile and must be returned by expensive methods, e.g. off gas purification. This concerns especially the active compounds of cesium and ruthenium. The heat conductivity of the glass matrix is relatively small. In order not to exceed an impermissibly high central temperature of the cask, caused by the residual heat, therefore waste concentration and block diameters are limited to values of about 20 weight % or 20 to 30 cm. Furthermore, there occur through the difference between the thermal coefficients of glass and container material mechanical stresses in the cooling which can lead to undesired stress corrosion and formation of fissures in the glass. The cooling off time necessary for waste casks can amount to several days in order to produce a fissure free package. Therefore this additional process step requires expensive hot cell space.
The bitumenization is usable only on relatively low activity concentrations, e.g. for the so-called liquid middle active waste having a .beta.,.gamma.-activity of about 0.1-1 Ci. Temperatures of 150.degree.-200.degree. C. are necessary which require expensive safety precautions, e.g. against burning. Besides bitumen under irradiation forms radiolysis gases such as, hydrogen.
The simple technique of cementation also has its disadvantages. Thus there are obtained with the same amounts of waste greater volumes of waste, e.g. compared to binding in bitumen 3 to 5 times the volume, a relatively low leach resistance of the product due to the porosity of the cement, and a radiolysis of the waste bound in the cement, which can lead to relatively large amounts of gas such as hydrogen.
For the embedding in polymerizable resins there are basically employed hydrocarbon compounds. Therefore the brittleness of the synthetic resin can be increased by the radiation effect of the radioactive waste and therewith the mechanical integrity of the package can be endangered. It is also true for such molded bodies that there is only a relatively slight resistance to radiolysis and the setting free of hydrogen.
There is known from German OS No. 2,756,700 a process for embeeding radioactive waste in a metal matrix which is formed by isostatically pressing around the waste with metal powder at a temperature between 1000.degree. and 1500.degree. C. The high pressing temperatures and the large consumption of corrosion resistant metal makes it appear that this process has little suitability at least for large bodies and for enclosing volatile radioactive materials.
Besides there are known so-called final storage containers which receive the waste material and which to obtain a sufficiently long interval of corrosion resistance for the most part are built as multilayer containers. As container materials there are used corrosion resistant metallic and non metallic materials.
In German OS No. 2,917,743 (and related U.S. Pat. No. 4,407,742, the entire disclosure of which is hereby incorporated by reference and relied upon) there is described a process in which radioactive and toxic wastes under careful conditions are bound into a good conducting carbon matrix which consists of a mixture of powdered carbon, preferably graphite, with a binding agent, whereby through pressing with the waste mixed in at a temperature above 100.degree. C. there is formed a corresponding molded body. As binding agent there can be employed organic and inorganic materials, the use of sulfur is advantageous and in a preferred embodiment there is employed a mixture of sulfur and nickel which forms the slightly water soluble nickel sulfide at a pressing temperature of about 400.degree. C. It is true that this matrix is corrosion and leach resitanst, but the waste materials can still be dissolved out of the surface layer.
For an economical use of this process there is required the highest possible concentration of waste in the molded body. However, at high portions of waste in the molded body the waste is leached out from ever deeper layers over a long period of time, and finally is leached out of the entire molded body.
Therefore it was the problem of the present invention to provide a molded body made of graphite and an inorganic binder for the safe, long time fixation of radioactive and toxic wastes, which has high density and is corrosion and leach resistant so that the embedded waste cannot be dissolved out even in very long periods of time.